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Simplify OpenMCSettings class (can properly delegate to openmc.Settings)

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This commit is contained in:
Paul Romano 2018-02-14 12:46:23 -06:00
parent dba919c610
commit 939d47cffa
3 changed files with 56 additions and 107 deletions
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116
openmc/deplete/openmc_wrapper.py
View file

@ -31,7 +31,7 @@ from .function import Settings, Operator
def chunks(items, n):
def _chunks(items, n):
min_size, extra = divmod(len(items), n)
j = 0
chunk_list = []
@ -43,70 +43,61 @@ def chunks(items, n):
class OpenMCSettings(Settings):
"""The OpenMCSettings class.
Extends Settings to provide information OpenMC needs to run.
"""Extends Settings to provide information OpenMC needs to run.
Attributes
----------
dt_vec : numpy.array
Array of time steps to take. (From Settings)
tol : float
Tolerance for adaptive time stepping. (From Settings)
Array of time steps to in units of [s]
output_dir : str
Path to output directory to save results. (From Settings)
Path to output directory to save results.
chain_file : str
Path to the depletion chain xml file. Defaults to the environment
variable "OPENDEPLETE_CHAIN" if it exists.
particles : int
Number of particles to simulate per batch.
batches : int
Number of batches.
inactive : int
Number of inactive batches.
lower_left : list of float
Coordinate of lower left of bounding box of geometry.
upper_right : list of float
Coordinate of upper right of bounding box of geometry.
entropy_dimension : list of int
Grid size of entropy.
dilute_initial : float, default 1.0e3
Path to the depletion chain xml file. Defaults to the
:envvar:`OPENDEPLETE_CHAIN` environment variable if it exists.
dilute_initial : float
Initial atom density to add for nuclides that are zero in initial
condition to ensure they exist in the decay chain. Only done for
nuclides with reaction rates.
nuclides with reaction rates. Defaults to 1.0e3.
round_number : bool
Whether or not to round output to OpenMC to 8 digits.
Useful in testing, as OpenMC is incredibly sensitive to exact values.
constant_seed : int
If present, all runs will be performed with this seed.
power : float
Power of the reactor in W. For a 2D problem, the power can be given in
Power of the reactor in [W]. For a 2D problem, the power can be given in
W/cm as long as the "volume" assigned to a depletion material is
actually an area in cm^2.
settings : openmc.Settings
Settings for OpenMC simulations
"""
_depletion_attrs = {'dt_vec', 'output_dir', 'chain_file', 'dilute_initial',
'round_number', 'power'}
def __init__(self):
super().__init__()
# OpenMC specific
try:
self.chain_file = os.environ["OPENDEPLETE_CHAIN"]
except KeyError:
self.chain_file = None
self.particles = None
self.batches = None
self.inactive = None
self.lower_left = None
self.upper_right = None
self.entropy_dimension = None
self.dilute_initial = 1.0e3
# OpenMC testing specific
self.round_number = False
self.constant_seed = None
# Depletion problem specific
self.power = None
# Avoid setattr to create OpenMC settings
self.__dict__['settings'] = openmc.Settings()
def __setattr__(self, name, value):
if name in self._depletion_attrs:
self.__dict__[name] = value
else:
setattr(self.__dict__['settings'], name, value)
def __getattr__(self, name):
if name in self._depletion_attrs:
return self.__dict__[name]
else:
return getattr(self.__dict__['settings'], name)
class Materials(object):
"""The Materials class.
@ -290,8 +281,8 @@ class OpenMCOperator(Operator):
i += 1
# Decompose geometry
mat_burn_lists = chunks(mat_burn, comm.size)
mat_not_burn_lists = chunks(mat_not_burn, comm.size)
mat_burn_lists = _chunks(mat_burn, comm.size)
mat_not_burn_lists = _chunks(mat_not_burn, comm.size)
mat_tally_ind = OrderedDict()
@ -456,7 +447,7 @@ class OpenMCOperator(Operator):
# Create XML files
if comm.rank == 0:
self.geometry.export_to_xml()
self.generate_settings_xml()
self.settings.settings.export_to_xml()
self.generate_materials_xml()
# Initialize OpenMC library
@ -526,46 +517,6 @@ class OpenMCOperator(Operator):
materials.export_to_xml()
def generate_settings_xml(self):
"""Generates settings.xml.
This function creates settings.xml using the value of the settings
variable.
Todo
----
Rewrite to generalize source box.
"""
batches = self.settings.batches
inactive = self.settings.inactive
particles = self.settings.particles
# Just a generic settings file to get it running.
settings_file = openmc.Settings()
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
settings_file.source = openmc.Source(space=openmc.stats.Box(
self.settings.lower_left, self.settings.upper_right))
if self.settings.entropy_dimension is not None:
entropy_mesh = openmc.Mesh()
entropy_mesh.lower_left = self.settings.lower_left
entropy_mesh.upper_right = self.settings.upper_right
entropy_mesh.dimension = self.settings.entropy_dimension
settings_file.entropy_mesh = entropy_mesh
# Set seed
if self.settings.constant_seed is not None:
seed = self.settings.constant_seed
else:
seed = random.randint(1, sys.maxsize-1)
settings_file.seed = self.seed = seed
settings_file.export_to_xml()
def _get_tally_nuclides(self):
nuc_set = set()
@ -581,7 +532,6 @@ class OpenMCOperator(Operator):
for i in range(1, comm.size):
nuc_newset = comm.recv(source=i, tag=i)
nuc_set |= nuc_newset
else:
comm.send(nuc_set, dest=0, tag=comm.rank)

18
scripts/example_run.py
View file

@ -1,6 +1,8 @@
"""An example file showing how to run a simulation."""
import numpy as np
import openmc
from openmc.data import JOULE_PER_EV
import openmc.deplete
import example_geometry
@ -15,20 +17,18 @@ N = np.floor(dt2/dt1)
dt = np.repeat([dt1], N)
# Create settings variable
# Depletion settings
settings = openmc.deplete.OpenMCSettings()
settings.power = 2.337e15*4*JOULE_PER_EV*1e6 # MeV/second cm from CASMO
settings.dt_vec = dt
settings.output_dir = 'test'
# OpenMC-delegated settings
settings.particles = 1000
settings.batches = 100
settings.inactive = 40
settings.lower_left = lower_left
settings.upper_right = upper_right
settings.entropy_dimension = [10, 10, 1]
joule_per_mev = 1.6021766208e-13
settings.power = 2.337e15*4*joule_per_mev # MeV/second cm from CASMO
settings.dt_vec = dt
settings.output_dir = 'test'
settings.source = openmc.Source(space=openmc.stats.Box(lower_left, upper_right))
settings.verbosity = 3
op = openmc.deplete.OpenMCOperator(geometry, settings)

29
tests/regression_tests/test_deplete_full.py
View file

@ -5,6 +5,8 @@ import shutil
from pathlib import Path
import numpy as np
import openmc
from openmc.data import JOULE_PER_EV
import openmc.deplete
from openmc.deplete import results
from openmc.deplete import utilities
@ -35,26 +37,23 @@ def test_full(run_in_tmpdir):
N = floor(dt2/dt1)
dt = np.full(N, dt1)
# Create settings variable
# Depletion settings
settings = openmc.deplete.OpenMCSettings()
settings.chain_file = str(Path(__file__).parents[2] / 'chains' /
'chain_simple.xml')
settings.power = 2.337e15*4*JOULE_PER_EV*1e6 # MeV/second cm from CASMO
settings.dt_vec = dt
settings.output_dir = "test_full"
settings.round_number = True
chain_file = str(Path(__file__).parents[2] / 'chains' / 'chain_simple.xml')
settings.chain_file = chain_file
# Add OpenMC-specific settings
settings.particles = 100
settings.batches = 100
settings.inactive = 40
settings.lower_left = lower_left
settings.upper_right = upper_right
settings.entropy_dimension = [10, 10, 1]
settings.round_number = True
settings.constant_seed = 1
joule_per_mev = 1.6021766208e-13
settings.power = 2.337e15*4*joule_per_mev # MeV/second cm from CASMO
settings.dt_vec = dt
settings.output_dir = "test_full"
space = openmc.stats.Box(lower_left, upper_right)
settings.source = openmc.Source(space=space)
settings.seed = 1
settings.verbosity = 3
op = openmc.deplete.OpenMCOperator(geometry, settings)